1. Field of the Invention
The subject invention is related to nuclear reactors as applied to the generation of electrical power and, more particularly, to protective systems for the reactor and mechanisms for containment of coolant accidentally lost from the system.
2. Description of the Prior Art
In pressurized water nuclear reactors, heat developed by a controlled nuclear reaction is used to generate steam. The steam drives turbine generators that produce electrical power. To control radioactive contamination in the system, the steam is generated indirectly by heat exchangers that are supplied on their primary side by pressurized water that is pumped through primary coolant loops that are connected to the reactor vessel.
Water is circulated through the coolant loops at extremely high temperature and pressure. In the event of a break in the coolant supply loop, water must continue to be circulated through the reactor coolant system to remove core decay heat as the reactor is shut down. Thus, a continuous flow of water would pass through the core and spill out of the break to the floor of the containment building.
In known reactor systems, the reactor vessel and coolant loops are housed in a large, reinforced concrete structure generally referred to as a containment building. In addition to protecting the reactor and the components of the system from external forces, the containment building provides a barrier to the spread of radioactive coolant in the event of an accidental loss of coolant.
If a substantial break were to develop in the pressurized water system, the loss of pressure and coolant would cause a "blowdown" in the reactor wherein the water in the reactor core would rapidly vaporize and coolant would be expelled from the reactor vessel. Thus, such a break would leave the reactor core uncovered and subject to potential damage. Accordingly, safety injection systems have been included in the prior systems whereby, in the event of an accidental loss of coolant, supplemental coolant is supplied to reflood the reactor. In some systems, a pressurized accumulator provides coolant in response to a drop in system pressure. In other systems such as described in U.S. Pat. No. 3,929,567, supplemental emergency water cooling tanks are provided at a level sufficiently higher than the reactor vessel so that coolant is passively provided to the reactor. Still other proposed systems such as in U.S. Pat. Nos. 4,050,983 and 4,210,614 incorporate a deluge tank that is accessed through a rupture disc upon an accidental loss of coolant.
In some systems, an emergency water supply was provided for reflooding the reactor core after the accumulator had refilled the lower plenum and downcomer of the reactor vessel. The emergency water supply was stored in an auxiliary building and provided to the reactor through a suction line and redundant safety injection pumps. In these systems, the pumps had to be reconnected to the containment building sump upon depletion of the water level in the emergency water storage tank. There was no diverse system to reflood the reactor core if the safety injection pumps should fail to operate. Furthermore, the injection pumps that were used in the core reflood operation were designed to operate at pressures far below the normal operating pressure for the reactor system.
In addition to the safety injection system, other auxiliary systems have also supported the normal and emergency operation of the nuclear power plant. Included among these auxiliary systems are containment spray systems, and the normal decay heat removal or residual heat removal system. Those systems were generally supplied by the same water storage tank. In some systems, such as described in U.S. Pat. No. 4,104,119, the auxiliary water storage tank is also used as a heat sink and for an emergency water supply for the steam generator.
The substantial cross-connection between auxiliary systems outside the containment building resulted in the location of all valves, pumps, heat exchangers and other associated equipment in an auxiliary building remotely located from the containment building. A loss of coolant accident or a normal decay heat removal operation required that radioactive water be pumped from the containment building into the auxiliary building. The auxiliary building was compartmentalized with respect to these systems and is designed according to Government regulations to contain a coolant leak at a specified rate for a limited period of time. However, if the leak were to exceed the specified rate or were to continue for a time in excess of the specified period, the potential existed for loss of containment of the leak. Furthermore, the auxiliary building tanks were insufficient to accommodate the volume of contaminated water that could be lost from the containment building. Therefore, there was also a potential that contaminated water would eventually flood the auxiliary building. As another complication, if the auxiliary building tanks were to rupture, there would be no source of water for the coolant injection system or auxiliary system and, therefore, nothing to prevent a core meltdown and containment rupture upon an accidental loss of coolant. Accordingly, there was a need to avoid circulation of radioactive water through the auxiliary building.
In systems such as disclosed in U.S. Pat. No. 4,092,490, the containment building is provided with extensions intended to protect the piping and valves adjacent the containment building. Other systems have attempted to avoid extension of the containment building by locating the auxiliary equipment between the inner liner and the outer shell of the containment building. One example of such a system is illustrated in U.S. Pat. No. 3,929,568. However, the emergency feed water is not confined to the containment building by either of these systems.
In all these prior art systems, the potential remained for escape of radioactive water outside the containment building, in the event of an accidental loss of coolant. Accordingly, there was a need in the prior art for a nuclear reactor system wherein the protection to the core associated with a loss of coolant accident would be improved and the potential for the escape of radioactive coolant outside containment significantly reduced.
Furthermore, the prior systems have not generally provided for an economically and practically feasible manner in which a nuclear power plant could promptly be returned to operation following a loss of coolant accident. In prior systems, the containment building had no special means to limit or control the spread of coolant within the building once it escaped from the coolant loop. Thus, the task of recovering any spilled coolant and cleaning up the reactor system was very time-consuming and expensive. As the refueling water storage tank was drained, the residual heat removal system, safety injection system, containment spray systems and other auxiliary systems were reconnected to the containment building sump as the water source. In this case, the containment spray system would spray highly radioactive fluid throughout the entire containment building. Re-entry of the containment building would thereafter require a remote decontamination of the entire building--including removal and reprocessing of perhaps 500,000 to 700,000 gallons of radioactive water. Accordingly, there was a need for a system that could be recovered and quickly and efficiently returned to service.
Another disadvantage with prior systems has been that there was no convenient mechanism to dynamically test the pumps and related valves and piping. For example, in some systems, performance could be verified only during refueling operations when the top of the reactor vessel was removed. Even then an elaborate and time-consuming procedure was required to develop a few data points from which the entire range of performace had to extrapolated. Therefore, there was also a need for an improved system that would permit the dynamic testing of the pumps and valves in an efficient manner.